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Reactivity Studies on the Advanced Neutron Source

Conference ·
OSTI ID:6491967
 [1];  [1];  [1]
  1. Idaho National Engineering Laboratory (INEL), Idaho Falls, ID (United States). EG&G Idaho, Inc.
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An Advanced Neutron Source (ANS) with a peak thermal neutron flux of about 8.5 x 1019 m-2s-1 is being designed for condensed matter physics, materials science, isotope production, and fundamental physics research. The ANS is a new reactor-based research facility being planned by Oak Ridge National Laboratory (ORNL) to meet the need for an intense steady-state source of neutrons. The design effort is currently in the conceptual phase. A reference reactor design has been selected in order to examine the safety, performance, and costs associated with this one design. The ANS Project has an established, documented safety philosophy, and safety-related design criteria are currently being established. The purpose of this paper is to present analyses of safety aspects of the reference reactor design that are related to core reactivity events. These analyses include control rod worth, shutdown rod worth, heavy water voiding, neutron beam tube flooding, light water ingress, and single fuel element criticality. Understanding these safety aspects will allow us to make design modifications that improve the reactor safety and achieve the safety related design criteria.
Research Organization:
Idaho National Engineering Laboratory (INEL), Idaho Falls, ID (United States). EG&G Idaho, Inc.
Sponsoring Organization:
USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
DOE Contract Number:
AC07-76ID01570
OSTI ID:
6491967
Report Number(s):
EGG-M-89519; CONF-900917--20; ON: DE91001916
Country of Publication:
United States
Language:
English

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Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
220100 -- Nuclear Reactor Technology-- Theory & Calculation
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
COMPUTER CODES
CONTROL ROD WORTHS
DESIGN
Evaluated Nuclear Data File (ENDF)
FUEL ELEMENTS
HEAVY WATER
HYDROGEN COMPOUNDS
Hafnium Control Rods
Heavy Water Reactor
M CODES
MCNP-3B
Monte Carlo Theory Models
NEUTRON SOURCE FACILITIES
Nuclear Criticality Safety Program (NCSP)
OXYGEN COMPOUNDS
PHYSICS
R CODES
REACTIVITY
REACTIVITY COEFFICIENTS
REACTOR COMPONENTS
REACTOR CORES
REACTOR PHYSICS
REACTOR SAFETY
REACTORS
RELAP-5
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SAFETY
Thermal Neutron Flux
VOID COEFFICIENT
WATER
WATER INFLUX